[Federal Register Volume 81, Number 148 (Tuesday, August 2, 2016)]
[Notices]
[Pages 50729-50742]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2016-18290]



[[Page 50729]]

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NUCLEAR REGULATORY COMMISSION

[NRC-2016-0151]


Biweekly Notice; Applications and Amendments to Facility 
Operating Licenses and Combined Licenses Involving No Significant 
Hazards Considerations

AGENCY: Nuclear Regulatory Commission.

ACTION: Biweekly notice.

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SUMMARY: Pursuant to Section 189a. (2) of the Atomic Energy Act of 
1954, as amended (the Act), the U.S. Nuclear Regulatory Commission 
(NRC) is publishing this regular biweekly notice. The Act requires the 
Commission to publish notice of any amendments issued, or proposed to 
be issued, and grants the Commission the authority to issue and make 
immediately effective any amendment to an operating license or combined 
license, as applicable, upon a determination by the Commission that 
such amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from July 5, 2019, to July 19, 2016. The last 
biweekly notice was published on July 19, 2016 (81 FR 46958).

DATES: Comments must be filed by September 1, 2016. A request for a 
hearing must be filed by October 3, 2016.

ADDRESSES: You may submit comments by any of the following methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2016-0151. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected]. For technical questions, contact 
the individual listed in the FOR FURTHER INFORMATION CONTACT section of 
this document.
     Mail comments to: Cindy Bladey, Office of Administration, 
Mail Stop: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington, 
DC 20555-0001.
    For additional direction on obtaining information and submitting 
comments, see ``Obtaining Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

FOR FURTHER INFORMATION CONTACT: Lynn Ronewicz, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 
20555-0001; telephone: 301-415-1927, email: [email protected].

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2016-0151, facility name, unit 
number(s), plant docket number, application date, and subject when 
contacting the NRC about the availability of information for this 
action. You may obtain publicly-available information related to this 
action by any of the following methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2016-0151.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and 
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, 
please contact the NRC's Public Document Room (PDR) reference staff at 
1-800-397-4209, 301-415-4737, or by email to [email protected]. The 
ADAMS accession number for each document referenced (if it is available 
in ADAMS) is provided the first time that it is mentioned in this 
document.
     NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

    Please include Docket ID NRC-2016-0151, facility name, unit 
number(s), plant docket number, application date, and subject in your 
comment submission.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment into ADAMS.
    I. Notice of Consideration of Issuance of Amendments to Facility 
Operating Licenses and Combined Licenses and Proposed No Significant 
Hazards Consideration Determination
    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in Sec.  50.92 of title 10 of the Code of 
Federal Regulations (10 CFR), this means that operation of the facility 
in accordance with the proposed amendment would not (1) involve a 
significant increase in the probability or consequences of an accident 
previously evaluated, or (2) create the possibility of a new or 
different kind of accident from any accident previously evaluated; or 
(3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period if circumstances change during the 30-day comment 
period such that failure to act in a timely way would result, for 
example in derating or shutdown of the facility. If the Commission 
takes action prior to the expiration of either the comment period or 
the notice period, it will publish in the Federal Register a notice of 
issuance. If the Commission makes a final no significant hazards 
consideration determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.

A. Opportunity To Request a Hearing and Petition for Leave To Intervene

    Within 60 days after the date of publication of this notice, any 
person(s) whose interest may be affected by this

[[Page 50730]]

action may file a request for a hearing and a petition to intervene 
with respect to issuance of the amendment to the subject facility 
operating license or combined license. Requests for a hearing and a 
petition for leave to intervene shall be filed in accordance with the 
Commission's ``Agency Rules of Practice and Procedure'' in 10 CFR part 
2. Interested person(s) should consult a current copy of 10 CFR 2.309, 
which is available at the NRC's PDR, located at One White Flint North, 
Room O1-F21, 11555 Rockville Pike (first floor), Rockville, Maryland 
20852. The NRC's regulations are accessible electronically from the NRC 
Library on the NRC's Web site at http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to 
intervene is filed within 60 days, the Commission or a presiding 
officer designated by the Commission or by the Chief Administrative 
Judge of the Atomic Safety and Licensing Board Panel, will rule on the 
request and/or petition; and the Secretary or the Chief Administrative 
Judge of the Atomic Safety and Licensing Board will issue a notice of a 
hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the requestor/petitioner 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
requestor/petitioner shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing. 
The requestor/petitioner must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
requestor/petitioner intends to rely to establish those facts or expert 
opinion to support its position on this issue. The petition must 
include sufficient information to show that a genuine dispute exists 
with the applicant on a material issue of law or fact. Contentions 
shall be limited to matters within the scope of the amendment under 
consideration. The contention must be one which, if proven, would 
entitle the requestor/petitioner to relief. A requestor/petitioner who 
fails to satisfy these requirements with respect to at least one 
contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing with respect to resolution of that person's admitted 
contentions, including the opportunity to present evidence and to 
submit a cross-examination plan for cross-examination of witnesses, 
consistent with the NRC's regulations, policies and procedures.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Requests for hearing, 
petitions for leave to intervene, and motions for leave to file new or 
amended contentions that are filed after the 60-day deadline will not 
be entertained absent a determination by the presiding officer that the 
filing demonstrates good cause by satisfying the three factors in 10 
CFR 2.309(c)(1)(i)-(iii). If a hearing is requested, and the Commission 
has not made a final determination on the issue of no significant 
hazards consideration, the Commission will make a final determination 
on the issue of no significant hazards consideration. The final 
determination will serve to decide when the hearing is held. If the 
final determination is that the amendment request involves no 
significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment. If the final determination is that the amendment 
request involves a significant hazards consideration, then any hearing 
held would take place before the issuance of any amendment unless the 
Commission finds an imminent danger to the health or safety of the 
public, in which case it will issue an appropriate order or rule under 
10 CFR part 2.
    A State, local governmental body, Federally-recognized Indian 
Tribe, or agency thereof, may submit a petition to the Commission to 
participate as a party under 10 CFR 2.309(h)(1). The petition should 
state the nature and extent of the petitioner's interest in the 
proceeding. The petition should be submitted to the Commission by 
September 19, 2016. The petition must be filed in accordance with the 
filing instructions in the ``Electronic Submissions (E-Filing)'' 
section of this document, and should meet the requirements for 
petitions for leave to intervene set forth in this section, except that 
under 10 CFR 2.309(h)(2) a State, local governmental body, or 
Federally-recognized Indian Tribe, or agency thereof does not need to 
address the standing requirements in 10 CFR 2.309(d) if the facility is 
located within its boundaries. A State, local governmental body, 
Federally-recognized Indian Tribe, or agency thereof may also have the 
opportunity to participate under 10 CFR 2.315(c).
    If a hearing is granted, any person who does not wish, or is not 
qualified, to become a party to the proceeding may, in the discretion 
of the presiding officer, be permitted to make a limited appearance 
pursuant to the provisions of 10 CFR 2.315(a). A person making a 
limited appearance may make an oral or written statement of position on 
the issues, but may not otherwise participate in the proceeding. A 
limited appearance may be made at any session of the hearing or at any 
prehearing conference, subject to the limits and conditions as may be 
imposed by the presiding officer. Details regarding the opportunity to 
make a limited appearance will be provided by the presiding officer if 
such sessions are scheduled.

B. Electronic Submissions (E-Filing)

    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing, a petition for leave to intervene, any motion or 
other document filed in the proceeding prior to the submission of a 
request for hearing or petition to intervene, and documents filed by 
interested governmental entities participating under 10 CFR 2.315(c), 
must be filed in accordance with the NRC's E-Filing rule (72 FR 49139; 
August 28, 2007, as amended at 77 FR 46562, August 3, 2012). The E-
Filing process requires participants to submit and serve all 
adjudicatory documents over the internet, or in some cases to mail 
copies on electronic storage media. Participants may not submit paper 
copies of their filings unless they seek an exemption in

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accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 10 
days prior to the filing deadline, the participant should contact the 
Office of the Secretary by email at [email protected], or by 
telephone at 301-415-1677, to request (1) a digital identification (ID) 
certificate, which allows the participant (or its counsel or 
representative) to digitally sign documents and access the E-Submittal 
server for any proceeding in which it is participating; and (2) advise 
the Secretary that the participant will be submitting a request or 
petition for hearing (even in instances in which the participant, or 
its counsel or representative, already holds an NRC-issued digital ID 
certificate). Based upon this information, the Secretary will establish 
an electronic docket for the hearing in this proceeding if the 
Secretary has not already established an electronic docket.
    Information about applying for a digital ID certificate is 
available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/getting-started.html. System requirements for accessing 
the E-Submittal server are detailed in the NRC's ``Guidance for 
Electronic Submission to the NRC,'' which is available on the agency's 
public Web site at http://www.nrc.gov/site-help/electronic-sub-ref-mat.html. Participants may attempt to use other software not listed on 
the Web site, but should note that the NRC's E-Filing system does not 
support unlisted software, and the NRC Electronic Filing Help Desk will 
not be able to offer assistance in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a request for 
hearing or petition for leave to intervene. Submissions should be in 
Portable Document Format (PDF) in accordance with NRC guidance 
available on the NRC's public Web site at http://www.nrc.gov/site-help/electronic-sub-ref-mat.html. A filing is considered complete at the 
time the documents are submitted through the NRC's E-Filing system. To 
be timely, an electronic filing must be submitted to the E-Filing 
system no later than 11:59 p.m. Eastern Time on the due date. Upon 
receipt of a transmission, the E-Filing system time-stamps the document 
and sends the submitter an email notice confirming receipt of the 
document. The E-Filing system also distributes an email notice that 
provides access to the document to the NRC's Office of the General 
Counsel and any others who have advised the Office of the Secretary 
that they wish to participate in the proceeding, so that the filer need 
not serve the documents on those participants separately. Therefore, 
applicants and other participants (or their counsel or representative) 
must apply for and receive a digital ID certificate before a hearing 
request/petition to intervene is filed so that they can obtain access 
to the document via the E-Filing system.
    A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC Electronic 
Filing Help Desk through the ``Contact Us'' link located on the NRC's 
public Web site at http://www.nrc.gov/site-help/e-submittals.html, by 
email to [email protected], or by a toll-free call at 1-866-672-
7640. The NRC Electronic Filing Help Desk is available between 9 a.m. 
and 7 p.m., Eastern Time, Monday through Friday, excluding government 
holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
stating why there is good cause for not filing electronically and 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, 
Maryland, 20852, Attention: Rulemaking and Adjudications Staff. 
Participants filing a document in this manner are responsible for 
serving the document on all other participants. Filing is considered 
complete by first-class mail as of the time of deposit in the mail, or 
by courier, express mail, or expedited delivery service upon depositing 
the document with the provider of the service. A presiding officer, 
having granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
http://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. 
However, in some instances, a hearing request and petition to intervene 
will require including information on local residence in order to 
demonstrate a proximity assertion of interest in the proceeding. With 
respect to copyrighted works, except for limited excerpts that serve 
the purpose of the adjudicatory filings and would constitute a Fair Use 
application, participants are requested not to include copyrighted 
materials in their submission.
    For further details with respect to these license amendment 
applications, see the application for amendment which is available for 
public inspection in ADAMS and at the NRC's PDR. For additional 
direction on obtaining information related to this document, see the 
``Obtaining Information and Submitting Comments'' section of this 
document.

Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station (CNS), Units 1 and 2, York County, South Carolina

    Date of amendment request: May 26, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16147A105.
    Description of amendment request: The amendments would revise 
Sections 8.3.1, ``AC Power Systems''; 9.2.1, ``Nuclear Service Water 
System''; 9.4.1, ``Control Room Area Ventilation''; and 9.4.3, 
``Auxiliary Building Ventilation System,'' of the updated final safety 
analysis report (UFSAR), to clarify how a shutdown unit supplying 
either its normal or emergency power source may be credited for 
operability of shared components supporting the operating unit.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?

[[Page 50732]]

    Response: No.
    The proposed change only involves a change to the UFSAR to 
reflect how shared systems at CNS can be powered from offsite or 
onsite power sources. The proposed change does not modify any plant 
equipment and does not impact any failure modes that could lead to 
an accident. Additionally, the proposed change does not impact the 
consequence of any analyzed accident since the change does not 
adversely affect any equipment related to accident mitigation.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change only involves a change to the UFSAR to 
reflect how shared systems at CNS can be powered from offsite or 
onsite power sources. The proposed change does not modify any plant 
equipment and there is no impact on the capability of the existing 
equipment to perform their intended functions. No system set points 
are being modified and no changes are being made to the method in 
which plant operations are conducted. No new failure modes are 
introduced by the proposed change and the proposed amendment does 
not introduce accident initiators or malfunctions that would cause a 
new or different kind of accident.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change only involves a change to the UFSAR to 
reflect how shared systems at CNS can be powered from offsite or 
onsite power sources. The proposed change to the UFSAR does not 
affect any of the assumptions used in the CNS accident analysis, nor 
does it affect any operability requirements for equipment important 
to safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Kate B. Nolan, Deputy General Counsel, Duke 
Energy Carolinas, LLC, 550 South Tryon Street--DEC45A, Charlotte, NC 
28202-1802.
    NRC Branch Chief: Michael T. Markley.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2, Beaver County, 
Pennsylvania

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346, 
Davis-Besse Nuclear Power Station, Unit No. 1, Ottawa County, Ohio

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, Perry 
Nuclear Power Plant, Unit No. 1, Lake County, Ohio

    Date of amendment request: May 24, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16148A047.
    Description of amendment request: The amendment would eliminate 
Technical Specification (TS), Section 5.5, ``Inservice Testing 
Program,'' to remove requirements duplicated in American Society of 
Mechanical Engineers (ASME) Code for Operations and Maintenance of 
Nuclear Power Plants (OM Code), Case OMN-20, ``Inservice Test 
Frequency.'' A new defined term, ``INSERVICE TESTING PROGRAM,'' will be 
added to TS Section 1.1, ``Definitions.'' The proposed change to the TS 
is consistent with TSTF-545, Revision 3, ``TS Inservice Testing Program 
Removal & Clarify SR Usage Rule Application to Section 5.5 Testing.''
    Using the consolidated line-item improvement process, the NRC staff 
issued a notice of availability in the Federal Register on March 28, 
2016 (81 FR 17208), for a possible proposed change that modifies the 
Standard Technical Specification (STS) to eliminate Chapter 5.0, 
``Administrative Controls,'' specification Section 5.5, ``Inservice 
Testing Program,'' to remove requirements duplicated in ASME Code, Case 
OMN-20, ``Inservice Test Frequency.'' ASME Code, Case OMN-20, provides 
similar definitions and allowances as in the current STS Inservice 
Testing Program. The notice of availability added a new defined term, 
``Inservice Testing Program (IST),'' to the STS, Section 1.1, 
``Definitions.'' Also, the STS, Section 3.0, ``Surveillance Requirement 
(SR) Applicability,'' and STS Bases were revised to explain the 
application of the usage rules to the Section 5.5 testing requirements. 
Existing uses of the term ``Inservice Testing Program'' in the STS and 
STS Bases were capitalized to indicate that it is now a defined term. 
The FR notice included the model application, No Significant Hazards 
Consideration (NSHC) Determination, and the model safety evaluation for 
referencing in license amendment applications. The licensee affirmed 
the applicability of the model NSHC determination in its application 
dated May 24, 2016, which is presented below.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below, along with NRC edits in square 
brackets:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises TS Chapter 5, ``Administrative 
Controls,'' Section 5.5, ``Programs and Manuals,'' by eliminating 
the ``Inservice Testing Program'' specification. Most requirements 
in the Inservice Testing Program are removed, as they are 
duplicative of requirements in the ASME OM Code, as clarified by 
Code Case OMN-20, ``Inservice Test Frequency.'' The remaining 
requirements in the Section 5.5 Inservice Testing Program are 
eliminated because the NRC has determined their inclusion in the TS 
is contrary to regulations. A new defined term, ``INSERVICE TESTING 
PROGRAM,'' is added to the TS, which references the requirements of 
10 CFR 50.55a(f).
    Performance of inservice testing is not an initiator to any 
accident previously evaluated. As a result, the probability of 
occurrence of an accident is not significantly affected by the 
proposed change. Inservice test frequencies under Code Case OMN-20 
are equivalent to the current testing period allowed by the TS with 
the exception that testing frequencies greater than 2 years may be 
extended by up to 6 months to facilitate test scheduling and 
consideration of plant operating conditions that may not be suitable 
for performance of the required testing. The testing frequency 
extension will not affect the ability of the components to mitigate 
any accident previously evaluated as the components are required to 
be operable during the testing period extension. Performance of 
inservice tests utilizing the allowances in OMN-20 will not 
significantly affect the reliability of the tested components. As a 
result, the availability of the affected components, as well as 
their ability to mitigate the consequences of accidents previously 
evaluated, is not affected.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any previously evaluated?
    Response: No.
    The proposed change does not alter the design or configuration 
of the plant. The proposed change does not involve a physical 
alteration of the plant; no new or different kind of equipment will 
be installed. The

[[Page 50733]]

proposed change does not alter the types of inservice testing 
performed. In most cases, the frequency of inservice testing is 
unchanged. However, the frequency of testing would not result in a 
new or different kind of accident from any previously evaluated 
since the testing methods are not altered.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No
    The proposed change eliminates some requirements from the TS in 
lieu of requirements in the ASME Code, as modified by use of Code 
Case OMN-20. Compliance with the ASME Code is required by 10 CFR 
50.55a. The proposed change also allows inservice tests with 
frequencies greater than 2 years to be extended by 6 months to 
facilitate test scheduling and consideration of plant operating 
conditions that may not be suitable for performance of the required 
testing. The testing frequency extension will not affect the ability 
of the components to respond to an accident as the components are 
required to be operable during the testing period extension. The 
proposed change will eliminate the existing TS SR 3.0.3 allowance to 
defer performance of missed inservice tests up to the duration of 
the specified testing frequency, and instead will require an 
assessment of the missed test on equipment operability. This 
assessment will consider the effect on a margin of safety (equipment 
operability). Should the component be inoperable, the Technical 
Specifications provide actions to ensure that the margin of safety 
is protected. The proposed change also eliminates a statement that 
nothing in the ASME Code should be construed to supersede the 
requirements of any TS. The NRC has determined that statement to be 
incorrect. However, elimination of the statement will have no effect 
on plant operation or safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, FirstEnergy Nuclear 
Operating Company, FirstEnergy Corporation, 76 South Main Street, Mail 
Stop A-GO-15, Akron, OH 44308.
    NRC Acting Branch Chief: G. Edward Miller.

Florida Power & Light Company, et al., Docket No. 50-389, St. Lucie 
Plant, Unit No. 2, St. Lucie County, Florida

    Date of amendment request: June 21, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16190A118.
    Description of amendment request: The amendment would update the 
Technical Specifications to revise the emergency diesel generator (EDG) 
engine-mounted fuel tank minimum volume from 200 gallons of fuel each 
to 238 gallons of fuel each.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The EDGs engine-mounted fuel oil tanks are part of a system used 
to mitigate the consequences of an accident and do not increase the 
probability of an accident previously evaluated. The increase in 
minimum fuel oil requirements enables operation of the EDGs to 
remain unchanged for ULSD [ultra low sulfur diesel] fuel oil, thus 
the EDGs continue to be capable of performing their design 
functions. Acceptance criteria continue to be satisfied. 
Accordingly, the proposed change does not increase the consequences 
of an accident.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    No new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the increase in 
minimum EDGs engine-mounted fuel oil tank volume. The proposed 
change has no adverse effect on any safety-related system and does 
not change the performance or integrity of any safety-related 
equipment. No new safety-related equipment is being added or 
replaced as a result of the proposed change.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The calculation for EDG fuel consumption shows that with the 
minimum day tank volume of 238 gallons of ULSD fuel, the requirement 
for two day tanks to provide a usable volume which is sufficient for 
at least 1 hour 100% load operation of one diesel generator set, 
plus a minimum margin of 10% is met. The day tank minimum volumes 
with the DOST [diesel oil storage tank] minimum volume is sufficient 
for the EDG loading increase due to potential operation at the upper 
frequency limit of 60.6 HZ [Hertz] (60 HZ, +1%) and the EPU 
[extended power uprate] requirements. The EDG fuel consumption 
analyses demonstrate that the EDG design continues to satisfy its 
safety function. The design basis limits for the accident and 
transient analyses will continue to meet their design criteria.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: William S. Blair, Managing Attorney--
Nuclear, Florida Power & Light Company, 700 Universe Boulevard, MS LAW/
JB, Juno Beach, FL 33408-0420.
    NRC Acting Branch Chief: Tracy J. Orf.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo 
Canyon Nuclear Power Plant, Units 1 and 2, San Luis Obispo County, 
California

    Date of amendment request: May 12, 2016. A publicly-available 
version is in ADAMS under Package Accession No. ML16146A100.
    Description of amendment request: The amendments would revise 
Technical Specification (TS) 5.5.6, ``Containment Leakage Rate Testing 
Program,'' to allow the following:
     Increase in the existing 10 CFR part 50, Appendix J, 
``Primary Reactor Containment Leakage Testing for Water-Cooled Power 
Reactors,'' Type A test interval from 10 years to 15 years in 
accordance with Nuclear Energy Institute (NEI) 94-01, Revision 2-A, 
``Industry Guideline for Implementing Performance-Based Option of 10 
CFR part 50, Appendix J,'' October 2008 (ADAMS Accession No. 
ML100620847).
     Adopt the use of American National Standards Institute/
American Nuclear Society (ANSI/ANS) 56.8-2002, ``Containment System 
Leakage Testing Requirements,'' as referenced in NEI 94-01, Revision 2-
A.
     Adopt an allowable test interval extension of 9 months, 
which is shorter than the currently allowed 25 percent grace, for the 
10 CFR 50, Appendix J, Type A, Type B, and Type C leakage tests in 
accordance with NEI 94-01, Revision 2-A.
    The proposed changes would revise TS 5.5.16 to replace the 
reference to NRC Regulatory Guide 1.163, ``Performance-Based 
Containment Leak-Test Program,'' September 1995 (ADAMS Accession No. 
ML003740058), and 10 CFR 50, Appendix J, Option B,

[[Page 50734]]

``Performance-Based Requirements,'' with a reference to NEI 94-01, 
Revision 2-A.
    In addition, the proposed amendments would modify TS 5.5.16 to 
remove an exception under paragraph 5.16.a.3 for a one-time 15-year 
Type A test interval beginning May 4, 1994, for Unit 1 and April 30, 
1993, for Unit 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed license amendment adopts the Nuclear Regulatory 
Commission (NRC)-accepted guidelines of Nuclear Energy Institute 
(NEI) Report 94-01, Revision 2-A, ``Industry Guideline for 
Implementing Performance-Based Option of 10 CFR part 50, Appendix 
J,'' for development of the Diablo Canyon Power Plant (DCPP) Units 1 
and 2 performance-based Technical Specification 5.5.16, 
``Containment Leakage Rate Testing Program.'' NEI 94-01 allows, 
based on risk and performance, an extension of Type A containment 
leak test intervals. Implementation of these guidelines continues to 
provide adequate assurance that during design basis accidents, the 
containment and its components will limit leakage rates to less than 
the values assumed in the plant safety analyses.
    The findings of the DCPP risk assessment confirm the general 
findings of previous studies that the risk impact with extending the 
containment leak rate is small, per the guidance provided in 
Regulatory Guide (RG) 1.174, Revision 2 ``An Approach for Using 
Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-
Specific Changes to the Licensing Basis,'' May 2011 (ADAMS Accession 
No. ML100910006).
    Since the license amendment is implementing a performance-based 
containment testing program, the proposed license amendment does not 
involve either a physical change to the plant or a change in the 
manner in which the plant is operated or controlled. The 
requirements for leakage rate tests and acceptance criteria will not 
be changed by this license amendment.
    Therefore, the containment will continue to perform its design 
function as a barrier to fission product releases.
    The proposed license amendment also deletes an exception 
previously granted to allow one time extensions of the Type A test 
frequency for DCPP. This exception was for an activity that has 
already taken place; therefore, the deletion is solely an 
administrative action that has no effect on any component and no 
physical impact on how the units are operated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different accident from any accident previously evaluated?
    Response: No.
    The proposed license amendment to implement a performance-based 
Type A testing program does not change the design or operation of 
structures, systems, or components of the plant. In addition, the 
proposed changes would not impact any other plant system or 
component.
    The proposed license amendment would continue to ensure 
containment integrity and would ensure operation within the bounds 
of existing accident analyses. There are no accident initiators 
created or affected by the proposed changes.
    The proposed license amendment also deletes an exception 
previously granted to allow one time extensions of the Type A test 
frequency for DCPP. This exception was for an activity that has 
already taken place; therefore, the deletion is solely an 
administrative action and does not change how the units are operated 
or maintained.
    Therefore, the proposed license amendment does not create the 
possibility of a new or different accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed license amendment to implement the performance-
based Type A testing program does not affect plant operations, 
design functions, or any analysis that verifies the capability of a 
structure, system, or component of the plant to perform a design 
function. In addition, this change does not affect safety limits, 
limiting safety system setpoints, or limiting conditions for 
operation.
    The specific requirements and conditions of Technical 
Specification 5.5.16, ``Containment Leakage Rate Testing Program,'' 
exist to ensure that the degree of containment structural integrity 
and leak-tightness that is considered in the plant safety analysis 
is maintained. The overall containment leak rate limit specified by 
the Technical Specifications is maintained. This ensures that the 
margin of safety in the plant safety analysis is maintained. The 
proposed amendment will ensure that the design, operation, testing 
methods and acceptance criteria for Type A tests specified in 
applicable codes and standards would continue to be met since these 
are not affected by implementation of a performance based Type A 
testing interval.
    The proposed amendment also deletes an exception previously 
granted to allow one time extensions of the Type A test frequency 
for DCPP. This exception was for an activity that has taken place; 
therefore, the deletion is solely an administrative action and does 
not change how the unit is operated and maintained.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jennifer Post, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, CA 94120.
    NRC Branch Chief: Robert J. Pascarelli.

South Carolina Electric and Gas Company and South Carolina Public 
Service Authority, Docket Nos. 52-027 and 52-028, Virgil C. Summer 
Nuclear Station, Units 2 and 3, Fairfield County, South Carolina

    Date of amendment request: June 16, 2016, as supplemented by letter 
dated July 7, 2016. Publicly-available versions are in ADAMS under 
Accession Nos. ML16168A282 and ML16189A453, respectively.
    Description of amendment request: The amendments propose changes to 
the Updated Final Safety Analysis Report (UFSAR) in the form of 
departures from the incorporated plant-specific Design Control Document 
Tier 2* and associated Tier 2 information. Specifically, the proposed 
departures consist of changes to the UFSAR to revise the details of the 
structural design of auxiliary building floors within module CA20 at 
approximate design elevations of 82'-6'' and 92'-6''.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The design functions of the auxiliary building floors are to 
provide support, protection, and separation for the seismic Category 
I mechanical and electrical equipment located in the auxiliary 
building. The auxiliary building is a seismic Category I structure 
and is designed for dead, live, thermal, pressure, safe shutdown 
earthquake loads, and loads due to postulated pipe breaks. The 
proposed changes to UFSAR descriptions are intended to address 
changes in the detail design of floors in the auxiliary building. 
The thickness and strength of the auxiliary building floors are not 
reduced. As a result, the design function of the auxiliary building 
structure is not adversely affected by the proposed changes. There 
is no change to plant systems or the response of systems to 
postulated accident conditions. There is no change to the predicted 
radioactive releases due to postulated accident conditions. The 
plant response to previously evaluated accidents or external events 
is not

[[Page 50735]]

adversely affected, nor do the changes described create any new 
accident precursors.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The changes to UFSAR descriptions are proposed to address 
changes in the detail design of floors in the auxiliary building. 
The thickness, geometry, and strength of the structures are not 
adversely altered. The concrete and reinforcement materials are not 
altered. The properties of the concrete are not altered. The changes 
to the design details of the auxiliary building structure do not 
create any new accident precursors. As a result, the design function 
of the auxiliary building structure is not adversely affected by the 
proposed changes.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The criteria and requirements of American Concrete Institute 
(ACI) 349 and American Institute of Steel Construction (AISC) N690 
provide a margin of safety to structural failure. The design of the 
auxiliary building structure conforms to criteria and requirements 
in ACI 349 and AISC N690 and therefore maintains the margin of 
safety. Analysis of the connection design confirms that code 
provisions are appropriate to the floor to wall connection. The 
proposed changes to the UFSAR address changes in the detail design 
of floors in the auxiliary building. The proposed changes also 
incorporate the requirements for development and anchoring of headed 
reinforcement which were previously approved. There is no change to 
design requirements of the auxiliary building structure. There is no 
change to the method of evaluation from that used in the design 
basis calculations. There is not a significant change to the in 
structure response spectra.
    Therefore, the proposed amendment does not result in a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP, 
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Acting Branch Chief: Jennifer Dixon-Herrity.

South Carolina Electric and Gas Company and South Carolina Public 
Service Authority, Docket Nos. 52-027 and 52-028, Virgil C. Summer 
Nuclear Station (VCSNS), Units 2 and 3, Fairfield County, South 
Carolina

    Date of amendment request: July 5, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16187A392.
    Description of amendment request: The amendment request relates to 
changes to the slab thickness between Column Lines I to J-1 and 2 to 4 
at plant elevation 153'-0''. The changes involve changes to 
incorporated AP1000 Design Control Document Tier 1 information and 
corresponding departures to Tier 2* Updated Final Safety Analysis 
Report information and conforming changes to the Combined License, 
Appendix C.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below, with NRC staff edits in square 
brackets:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The design functions of the nuclear island structures are to 
provide support, protection, and separation for the seismic Category 
I mechanical and electrical equipment located in the nuclear island. 
The nuclear island structures are structurally designed to meet 
seismic Category I requirements as defined in Regulatory Guide 1.29. 
The change of the thickness of the floor above the [Component 
Cooling Water System (CCS)] Valve room in the auxiliary building 
meets criteria and requirements of American Concrete Institute (ACI) 
349 and American Institute of Steel Construction (AISC) N690, does 
not have an adverse impact on the response of the nuclear island 
structures to safe shutdown earthquake ground motions or loads due 
to anticipated transients or postulated accident conditions. The 
proposed changes do not impact the support, design, or operation of 
mechanical and fluid systems. There is no change to plant systems or 
the response of systems to postulated accident conditions. There is 
no change to the predicted radioactive releases due to normal 
operation or postulated accident conditions. The plant response to 
previously evaluated accidents or external events is not adversely 
affected, nor does the change described create any new accident 
precursors.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change is to revise the thickness of the floor 
above the CCS Valve room in the auxiliary building. The proposed 
changes do not change the design requirements of the nuclear island 
structures. The proposed changes do not change the design function, 
support, design, or operation of mechanical and fluid systems. The 
proposed changes do not result in a new failure mechanism for the 
nuclear island structures or new accident precursors. As a result, 
the design function of the nuclear island structures is not 
adversely affected by the proposed change.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    No safety analysis or design basis acceptance limit/criterion is 
challenged or exceeded by the proposed changes, thus, no margin of 
safety is reduced.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety previously evaluated.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Kathryn M. Sutton, Morgan, Lewis & 
Bockius LLC, 1111 Pennsylvania Avenue NW., Washington, DC 20004-2514.
    Acting NRC Branch Chief: Jennifer Dixon-Herrity.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego 
County, California

    Date of amendment request: June 16, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16172A075.
    Description of amendment request: The amendments would extend the 
scheduled implementation date for Milestone 8 of the San Onofre Nuclear 
Generating Station, Units 2 and 3, Cyber Security Plan to December 31, 
2019, in order to more fully reflect the permanent shutdown status of 
the facility and accommodate ongoing decommissioning activities.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or

[[Page 50736]]

consequences of an accident previously evaluated?
    Response: No.
    The proposed change to the San Onofre Nuclear Generating Station 
(SONGS) Cyber Security Plan Implementation Schedule is 
administrative in nature. This change does not alter accident 
analysis assumptions, add any initiators, or affect the function of 
plant systems or the manner in which systems are operated, 
maintained, modified, tested, or inspected. The proposed change does 
not require any plant modifications which affect the performance 
capability of the structures, systems, and components (SSCs) relied 
upon to mitigate the consequences of postulated accidents, and has 
no impact on the probability or consequences of an accident 
previously evaluated.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change to the SONGS Cyber Security Plan 
Implementation Schedule is administrative in nature. This proposed 
change does not alter accident analysis assumptions, add any 
initiators, or affect the function of plant systems or the manner in 
which systems are operated, maintained, modified, tested, or 
inspected. The proposed change does not require any plant 
modifications which affect the performance capability of the SSCs 
relied upon to mitigate the consequences of postulated accidents, 
and does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Plant safety margins are established through limiting conditions 
for operation, limiting safety system settings, and safety limits 
specified in the technical specifications. The proposed change to 
the SONGS Cyber Security Plan Implementation Schedule is 
administrative in nature. Since the proposed change is 
administrative in nature, there is no change to these established 
safety margins.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Walker A. Matthews, Esquire, Southern 
California Edison Company, 2244 Walnut Grove Avenue, Rosemead, CA 
91770.
    NRC Branch Chief: Bruce Watson.

Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, 
Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County, 
Georgia

    Date of amendment request: March 4, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16064A352.
    Description of amendment request: The amendment proposes to change 
the VEGP, Units 3 and 4, License Conditions 2.D(12)(d) and submits the 
new plant-specific Emergency Action Level (EAL) scheme for both units.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The requested amendment proposes changes to the Vogtle Electric 
Generating Plant (VEGP) Units 3 and 4 License Conditions 2.D(12)(d) 
and submits the new plant-specific Emergency Action Level (EAL) 
scheme for both units. The proposed changes, including the 
modification of VEGP Units 3 and 4 License Condition 2.D(12)(d) and 
submittal of the new plant-specific EALs for both units, do not 
impact the physical function of plant structures, systems, or 
components (SSCs) or the manner in which SSCs perform their design 
function. The proposed changes neither adversely affect accident 
initiators or precursors, nor alter design assumptions. The proposed 
changes do not alter or prevent the ability of SSCs to perform their 
intended function to mitigate the consequences of an initiating 
event within assumed acceptance limits. No operating procedures or 
administrative controls that function to prevent or mitigate 
accidents are affected by the proposed changes.
    Therefore, the requested amendment does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes, including the modification of VEGP Units 3 
and 4 License Conditions 2.D(12)(d) and submittal of the new plant-
specific EALs for both units, do not involve a physical alteration 
of the plant (i.e., no new or different type of equipment will be 
installed or removed) or a change in the method of plant operation. 
The proposed changes will not introduce failure modes that could 
result in a new accident, and the changes do not alter assumptions 
made in the safety analysis. The proposed changes are not initiators 
of any accidents.
    Therefore, the requested amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    Margin of safety is associated with the ability of the fission 
product barriers (i.e., fuel cladding, reactor coolant system 
pressure boundary, and containment structure) to limit the level of 
radiation dose to the public. The proposed changes to the plant-
specific EALs and the modification of VEGP Units 3 and 4 License 
Conditions 2.D(12)(d) do not impact operation of the plant or its 
response to transients or accidents. The proposed changes do not 
affect the Technical Specifications. The proposed changes do not 
involve a change in the method of plant operation, and no accident 
analyses will be affected by the proposed changes.
    Additionally, the proposed changes will not relax any criteria 
used to establish safety limits and will not relax any safety system 
settings. The safety analysis acceptance criteria are not affected 
by these proposed changes. The proposed changes will not result in 
plant operation in a configuration outside the design basis. The 
proposed changes do not adversely affect systems that respond to 
safely shut down the plant and to maintain the plant in a safe 
shutdown condition.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. M. Stanford Blanton, Balch & Bingham 
LLP, 1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Acting Branch Chief: Jennifer Dixon-Herrity.

Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, 
Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County, 
Georgia

    Date of amendment request: April 26, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16117A531.
    Description of amendment request: The amendments would change the 
certified AP1000 Design Control Document (DCD) Tier 1 information and 
depart from the plant-specific Tier 2 and Tier 2* information in the 
Updated Final Safety Analysis Report (UFSAR) for VEGP, Units 3 and 4, 
by modifying the overall design of the Central Chilled Water subsystem 
to relocate the Air Cooled Chiller Pump 3 (VWS-MP-03)

[[Page 50737]]

and associated equipment from the Auxiliary Building to the Annex 
Building, for each unit respectively. The proposed changes include 
information in the Combined License, Appendix C. An exemption request 
relating to the proposed changes to the AP1000 DCD Tier 1 is included 
with the request.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The Central Chilled Water System (VWS) performs the nonsafety-
related function of supplying chilled water to the heating, 
ventilation, and air conditioning (HVAC) systems. The only safety-
related function of the VWS is to provide isolation of the VWS lines 
penetrating the containment. The low capacity VWS subsystem is non-
seismically designed. The change to relocate an air cooled chiller 
pump and associated equipment and add a chemical feed tank to this 
pump does not adversely affect the capability of either low capacity 
VWS subsystem loop to perform the system design function. This 
change does not have an adverse impact on the response to 
anticipated transient or postulated accident conditions because the 
low capacity VWS subsystem is a nonsafety-related and non-seismic 
system. No safety-related structure, system, component (SSC) or 
function is involved with or affected by this change. The changes to 
the low capacity VWS subsystem do not involve an interface with any 
SSC accident initiator or initiating sequence of events, and thus, 
the probabilities of the accidents evaluated in the plant-specific 
UFSAR [Updated Final Safety Analysis Report] are not affected. The 
proposed VWS change does not involve a change to the predicted 
radiological releases due to postulated accident conditions, thus, 
the consequences of the accidents evaluated in the UFSAR are not 
affected.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes to the nonsafety-related low capacity VWS 
subsystem do not affect any safety-related equipment, nor do they 
add any new interfaces to safety-related SSCs. No system or design 
function or equipment qualification is affected by these changes. 
The changes do not introduce a new failure mode, malfunction or 
sequence of events that could affect safety related equipment.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The VWS is a nonsafety-related system that performs the defense-
in-depth function of providing a reliable source of chilled water to 
various HVAC subsystems and unit coolers and the safety-related 
function of providing isolation of the VWS lines penetrating the 
containment. The changes to the VWS do not affect the VWS 
containment penetrations or any other safety related equipment or 
fission product barriers. The requested changes will not affect any 
design code, function, design analysis, safety analysis input or 
result, or design/safety margin. No safety analysis or design basis 
acceptance limit/criterion is challenged or exceeded by the 
requested changes.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP, 
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Acting Branch Chief: Jennifer Dixon-Herrity.

Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, 
Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia

    Date of amendment request: May 27, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16148A631.
    Description of amendment request: The amendment request proposes 
changes to the Combined License (COL), Appendix A, Technical 
Specifications (TSs), and Updated Final Safety Analysis Report (UFSAR) 
in the form of departures from the incorporated plant-specific Design 
Control Document Tier 2 information. Specifically, the proposed 
departures consist of changes to the UFSAR adding compensation for 
changes in reactor coolant density using the ``delta T'' power signal 
to the reactor coolant flow input signal for the low reactor coolant 
flow trip function of the Reactor Trip System (RTS). Additionally, TS 
Surveillance Requirement (SR) 3.3.1.3 is added to the surveillances 
required for the Reactor Coolant Flow[middot]Low reactor trip in TS 
Table 3.3.1-1, Function 7.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change adds compensation, for changes in reactor 
coolant density using the [delta T] power signal, to the reactor 
coolant flow input signal for the low reactor coolant flow reactor 
trip function of the RTS. The proposed change also adds TS SR 
3.3.1.3 to the surveillances required for the Reactor Coolant Flow-
Low reactor trip specified in TS Table 3.3.1-1. SR 3.3.1.3 compares 
the calorimetric heat balance to the calculated [delta T] power in 
each Protection and Safety Monitoring System (PMS) division every 24 
hours to assure acceptable [delta T] power calibration. As such, the 
surveillance is also required to support operability of the Reactor 
Coolant Flow-Low trip function. This change to the low reactor 
coolant flow trip input signal assures that the reactor will trip on 
low reactor coolant flow when the requisite conditions are met, and 
minimize spurious reactor trips and the accompanying plant 
transients. The change to the COL Appendix A Table 3.3.1-1 aligns 
the surveillance of the Reactor Coolant Flow-Low trip with the 
addition of the compensation, for changes in reactor coolant density 
using [delta T] power to the flow input signal to the trip. These 
changes do not affect the operation of any systems or equipment that 
initiate an analyzed accident or alter any structures, systems, and 
components (SSC) accident initiator or initiating sequence of 
events.
    These changes have no adverse impact on the support, design, or 
operation of mechanical and fluid systems. The response of systems 
to postulated accident conditions is not adversely affected and 
remains within response time assumed in the accident analysis. There 
is no change to the predicted radioactive releases due to normal 
operation or postulated accident conditions. Consequently, the plant 
response to previously evaluated accidents or external events is not 
adversely affected, nor does the proposed change create any new 
accident precursors.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes do not affect the operation of any systems 
or equipment that may initiate a new or different kind of accident, 
or alter any SSC such that a new accident initiator or initiating 
sequence of events is created. The proposed change adds 
compensation, for changes in reactor coolant density using [delta T] 
power signal, to the reactor coolant flow input signal to the low 
reactor coolant flow reactor trip function of the RTS. The proposed 
change also adds TS

[[Page 50738]]

SR 3.3.1.3 to the surveillances required for the Reactor Coolant 
Flow-Low reactor trip specified in TS Table 3.3.1-1. SR 3.3.1.3 
compares the calorimetric heat balance to the calculated [delta T] 
power in each PMS division every 24 hours to assure acceptable 
[delta T] power calibration. As such, the surveillance is also 
required to support operability of the Reactor Coolant Flow-Low trip 
function. The proposed change to the low reactor coolant flow 
reactor trip input signal does not alter the design function of the 
low flow reactor trip. The change to the COL Appendix A Table 3.3.1-
1 aligns the surveillance of the Reactor Coolant Flow-Low trip with 
the addition of compensation, for changes in reactor coolant density 
using [delta T] power to the flow input signal to the trip. 
Consequently, because the low reactor coolant flow trip functions 
are unchanged, there are no adverse effects that could create the 
possibility of a new or different kind of accident from any 
previously evaluated in the UFSAR.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    4. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change adds compensation, for changes in reactor 
coolant density using [delta T] power signal, to the reactor coolant 
flow input signal for the low reactor coolant flow trip function of 
the RTS. The proposed change also adds TS SR 3.3.1.3 to the 
surveillances required for the Reactor Coolant Flow-Low reactor trip 
specified in TS Table 3.3.1-1. SR 3.3.1.3 compares the calorimetric 
heat balance to the calculated [delta T] power in each PMS division 
every 24 hours to assure acceptable [delta T] power calibration. As 
such, the surveillance is also required to support operability of 
the Reactor Coolant Flow-Low trip function. The proposed changes do 
not alter any applicable design codes, code compliance, design 
function, or safety analysis. Consequently, no safety analysis or 
design basis acceptance limit/criterion is challenged or exceeded by 
the proposed change, thus the margin of safety is not reduced.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP, 
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Acting Branch Chief: Jennifer Dixon-Herrity.

Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, 
Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia

    Date of amendment request: June 14, 2016, as supplemented by letter 
dated July 1, 2016. Publicly-available versions are in ADAMS under 
Accession Nos. ML16166A409 and ML16183A394, respectively.
    Description of amendment request: The amendment request proposes 
changes to the Updated Final Safety Analysis Report (UFSAR) in the form 
of departures from the incorporated plant-specific Design Control 
Document Tier 2* and associated Tier 2 information. Specifically, the 
proposed departures consist of changes to the UFSAR to revise the 
details of the structural design of auxiliary building floors within 
module CA20 at approximate design elevations of 82'-6'' and 92'-6''.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The design functions of the auxiliary building floors are to 
provide support, protection, and separation for the seismic Category 
I mechanical and electrical equipment located in the auxiliary 
building. The auxiliary building is a seismic Category I structure 
and is designed for dead, live, thermal, pressure, safe shutdown 
earthquake loads, and loads due to postulated pipe breaks. The 
proposed changes to UFSAR descriptions are intended to address 
changes in the detail design of floors in the auxiliary building. 
The thickness and strength of the auxiliary building floors are not 
reduced. As a result, the design function of the auxiliary building 
structure is not adversely affected by the proposed changes. There 
is no change to plant systems or the response of systems to 
postulated accident conditions. There is no change to the predicted 
radioactive releases due to postulated accident conditions. The 
plant response to previously evaluated accidents or external events 
is not adversely affected, nor do the changes described create any 
new accident precursors. Therefore, the proposed amendment does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The changes to UFSAR descriptions are proposed to address 
changes in the detail design of floors in the auxiliary building. 
The thickness, geometry, and strength of the structures are not 
adversely altered. The concrete and reinforcement materials are not 
altered. The properties of the concrete are not altered. The changes 
to the design details of the auxiliary building structure do not 
create any new accident precursors. As a result, the design function 
of the auxiliary building structure is not adversely affected by the 
proposed changes.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The criteria and requirements of American Concrete Institute 
(ACI) 349 and American Institute of Steel Construction (AISC) N690 
provide a margin of safety to structural failure. The design of the 
auxiliary building structure conforms to criteria and requirements 
in ACI 349 and AISC N690 and therefore maintains the margin of 
safety. Analysis of the connection design confirms that code 
provisions are appropriate to the floor to wall connection. The 
proposed changes to the UFSAR address changes in the detail design 
of floors in the auxiliary building. The proposed changes also 
incorporate the requirements for development and anchoring of headed 
reinforcement which were previously approved. There is no change to 
design requirements of the auxiliary building structure. There is no 
change to the method of evaluation from that used in the design 
basis calculations. There is not a significant change to the in 
structure response spectra.
    Therefore, the proposed amendment does not result in a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP, 
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Acting Branch Chief: Jennifer Dixon-Herrity.

Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, 
Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia

    Date of amendment request: June 3, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16155A366.
    Description of amendment request: The amendment request proposes 
changes to correct editorial errors in Combined License (COL) Appendix 
C (and plant-specific Tier 1) and promote consistency with the Updated 
Final Safety Analysis Report (UFSAR) Tier 2

[[Page 50739]]

information. Additionally, one of the proposed changes to plant-
specific Tier 1 information also requires an involved change to UFSAR 
Tier 2 information. Pursuant to the provisions of 10 CFR 52.63(b)(1), 
an exemption from elements of the design as certified in the 10 CFR 
part 52, Appendix D, design certification rule is also requested for 
the plant-specific Tier 1 material departures. The requested amendment 
also contains a proposed editorial correction to COL paragraph 2.D.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed consistency and editorial Combined License (COL) 
Appendix C (and plant-specific Tier 1) and involved Tier 2 changes, 
along with one COL paragraph 2.D change, do not involve a technical 
change, (e.g. there is no design parameter or requirement, 
calculation, analysis, function or qualification change). No 
structure, system, component design or function would be affected. 
No design or safety analysis would be affected. The proposed changes 
do not affect any accident initiating event or component failure, 
thus the probabilities of the accidents previously evaluated are not 
affected. No function used to mitigate a radioactive material 
release and no radioactive material release source term is involved, 
thus the radiological releases in the accident analyses are not 
affected.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed consistency and editorial COL Appendix C (and 
plant-specific Tier 1) and involved Tier 2 changes, along with one 
COL paragraph 2.D change, would not affect the design or function of 
any structure, system, component (SSC), but will instead provide 
consistency between the SSC designs and functions currently 
presented in the Updated Final Safety Analysis Report (UFSAR) and 
the Tier 1 information. The proposed changes would not introduce a 
new failure mode, fault or sequence of events that could result in a 
radioactive material release.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed consistency and editorial COL Appendix C (and 
plant-specific Tier 1) and involved Tier 2 update, along with one 
COL paragraph 2.D change, is non-technical, thus would not affect 
any design parameter, function or analysis. There would be no change 
to an existing design basis, design function, regulatory criterion, 
or analysis. No safety analysis or design basis acceptance limit/
criterion is involved.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP, 
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
    NRC Acting Branch Chief: Jennifer Dixon-Herrity.

Tennessee Valley Authority Docket Nos. 50-259, 50-260, and 50-296, 
Browns Ferry Nuclear Plant (BFN), Unit 1, 2 and 3, Limestone County 
Alabama

Tennessee Valley Authority (TVA), Docket Nos. 50-327 and 50-328, 
Sequoyah Nuclear Plant (SQN), Units 1 and 2, Hamilton County, Tennessee

    Date of amendment request: April 14, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16105A287.
    Description of amendment request: The amendments would revise the 
BFN Units 1, 2, and 3, and the SQN, Units 1 and 2, Technical 
Specification (TS) 5.3, ``Unit Staff Qualifications,'' to delete the 
references to Regulatory Guide 1.8, Revision 2, and replace it with 
references to the TVA Nuclear Quality Assurance Plan (NQAP). The 
proposed changes would ensure consistent regulatory requirements 
regarding staff qualifications for the TVA nuclear fleet. The proposed 
changes would further allow TVA to implement standard procedures 
related to staff qualifications. Additionally, the proposed TS changes 
are consistent with the intent of NRC Administrative Letter 95-06 in 
that the relocated requirements are adequately controlled by 10 CFR 50, 
Appendix B, and the quality assurance change control process in 10 CFR 
50.54(a).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequence of an accident previously evaluated?
    Response: No.
    The Unit Staff Qualifications that are being removed from BFN TS 
5.3.1 and SQN TS 5.3.1 are redundant to requirements contained in 
Appendix B to the TVA NQAP and are consistent with the Watts Bar 
(WBN) Unit 1 and Unit 2 Technical Specifications (TS). Changes to 
the TVA NQAP are controlled by 10 CFR 50.54(a). These changes do not 
affect any of the design basis accidents.
    Therefore, the proposed changes do not involve an increase in 
the probability or consequences of an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The Unit Staff Qualifications that are being removed from BFN TS 
5.3.1 and SQN TS 5.3.1 are redundant to requirements contained in 
Appendix B to the TVA NQAP and are consistent with the WBN Unit 1 
and Unit 2 TS. Changes to the TVA NQAP are controlled by 10 CFR 
50.54(a). These changes do not affect any of the design basis 
accidents. No modifications to any plant equipment are involved. 
There is no effect on system interactions made by these changes.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The Unit Staff Qualifications that are being removed from BFN TS 
5.3.1 and SQN TS 5.3.1 are redundant to requirements contained in 
Appendix B to the TVA NQAP and are consistent with the WBN Unit 1 
and Unit 2 TS. Changes to the TVA NQAP are controlled by 10 CFR 
50.54(a). The margin of safety as reported in the basis for the TS 
is not reduced.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902.
    NRC Acting Branch Chief: Tracy J. Orf.

[[Page 50740]]

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant (SQN), Units 1 and 2, Hamilton County, Tennessee

    Date of amendment request: May 26, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16148A175.
    Description of amendment request: The amendments would modify the 
SQN, Units 1 and 2, Technical Specification (TS) 3.8.1, ``AC 
[Alternating Current] Sources--Operating,'' by revising the acceptance 
criteria for the diesel generator (DG) steady-state frequency 
acceptance criteria specified in the TS Surveillance Requirements 
(SRs). The frequency would be changed to address the non-conservative 
TS recently identified.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequence of an accident previously evaluated?
    Response: No.
    The DGs are required to be operable in the event of a design 
basis accident coincident with a loss of offsite power to mitigate 
the consequences of the accident. The DGs are not accident 
initiators and, therefore, these changes do not involve a 
significant increase in the probability of an accident previously 
evaluated.
    The accident analyses assume that at least the boards in one 
load group are provided with power either from the offsite circuits 
or the DGs. The change proposed in this license amendment request 
will continue to assure that the DGs have the capacity and 
capability to assume their maximum design basis accident loads. The 
proposed change does not significantly alter how the plant would 
mitigate an accident previously evaluated.
    The proposed change does not adversely affect accident 
initiators or precursors nor alter the design assumptions, 
conditions, and configuration of the facility or the manner in which 
the plant is operated and maintained. The proposed change does not 
adversely affect the ability of structures, systems, and components 
(SSC) to perform their intended safety function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The proposed change does not affect the source term, 
containment isolation, or radiological release assumptions used in 
evaluating the radiological consequences of any accident previously 
evaluated. Further, the proposed change does not increase the types 
and amounts of radioactive effluent that may be released offsite, 
nor significantly increase individual or cumulative occupational/
public radiation exposure.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not involve a change in the plant 
design, system operation, or the use of the DGs. The proposed change 
requires the DGs to meet SR acceptance criteria that envelope the 
actual demand requirements for the DGs during design basis 
conditions. These revised acceptance criteria continue to 
demonstrate the capability and capacity of the DGs to perform their 
required functions. There are no new failure modes or mechanisms 
created due to testing the DGs within the proposed acceptance 
criteria. Testing of the DGs at the proposed acceptance criteria 
does not involve any modification in the operational limits or 
physical design of plant systems. There are no new accident 
precursors generated due to the proposed test loadings.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change will continue to demonstrate that the DGs 
meet the TS definition of operability, that is, the proposed 
acceptance criteria will continue to demonstrate that the DGs will 
perform their safety function. The proposed testing will also 
continue to demonstrate the capability and capacity of the DGs to 
supply their required loads for mitigating a design basis accident.
    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings or limiting conditions for 
operation are determined. The safety analysis acceptance criteria 
are not affected by this change. The proposed change will not result 
in plant operation in a configuration outside the design basis.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, 6A West Tower, Knoxville, TN 37902.
    NRC Acting Branch Chief: Tracy J. Orf.

Tennessee Valley Authority, Docket Nos. 50-390 and 50-391, Watts Bar 
Nuclear Plant (WBN), Units 1 and 2, Rhea County, Tennessee

    Date of amendment request: June 7, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16159A208.
    Description of amendment request: The amendments would revise the 
WBN, Unit 2, Technical Specification (TS) 3.7.10, ``Control Room 
Emergency Ventilation System (CREVS),'' to include specific shutdown 
Required Actions and associated Completion Times during conditions to 
be taken due to a tornado warning. The proposed TS changes would be 
consistent with the current TS 3.7.10 for WBN, Unit 1. Additionally, 
the amendments would revise several administrative-related 
inconsistencies identified in the WBN, Units 1 and 2, TSs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes modify WBN Unit 1 TS 3.7.10 to resolve a 
potential conflict in applying the appropriate actions for not 
meeting the Required Action and associated Completion Time of 
Condition E and request administrative changes to correct 
inconsistencies in TS Applicability statements.
    The proposed changes do not affect the structures, systems, or 
components (SSCs) of the plant, affect plant operations, or any 
design function or an analysis that verifies the capability of an 
SSC to perform a design function. No change is being made to any of 
the previously evaluated accidents in the WBN Unit 1 Updated Final 
Safety Analysis Report (UFSAR) and the WBN Unit 2 FSAR [Final Safety 
Analysis Report]. These proposed changes are administrative or 
provide specific shutdown actions instead of using default shutdown 
actions.
    The proposed changes do not (1) require physical changes to 
plant systems, structures, or components; (2) prevent the safety 
function of any safety-related system, structure, or component 
during a design basis event; (3) alter, degrade, or prevent action 
described or assumed in any accident described in the WBN Unit 1 
UFSAR and the WBN Unit 2 FSAR from being perform[ed] because the 
safety-related systems, structures, or components are not modified; 
(4) alter any assumptions previously made in evaluating radiological 
consequences; or (5) affect the integrity of any fission product 
barrier.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?

[[Page 50741]]

    Response: No.
    The proposed changes do not introduce any new accident causal 
mechanisms, since no physical changes are being made to the plant, 
nor do they impact any plant systems that are potential accident 
initiators.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The margin of safety associated with the acceptance criteria of 
any accident is unchanged. The proposed changes will have no effect 
on the availability, operability, or performance of safety-related 
systems and components. The proposed change will not adversely 
affect the operation of plant equipment or the function of equipment 
assumed in the accident analysis.
    The proposed amendment does not involve changes to any safety 
analyses assumptions, safety limits, or limiting safety system 
settings. The changes do not adversely affect plant-operating 
margins or the reliability of equipment credited in the safety 
analyses.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Sherry Quirk, Executive Vice President and 
General Counsel, Tennessee Valley Authority, 400 West Summit Hill Dr., 
6A West Tower, Knoxville, TN 37902.
    NRC Acting Branch Chief: Tracy J. Orf.

III. Previously Published Notices of Consideration of Issuance of 
Amendments to Facility Operating Licenses and Combined Licenses, 
Proposed No Significant Hazards Consideration Determination, and 
Opportunity for a Hearing

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

Duke Energy Progress, Inc., Docket No. 50-400, Shearon Harris Nuclear 
Power Plant, Unit 1, Wake and Chatham Counties, North Carolina

    Date of amendment request: August 18, 2015, as supplemented by 
letters dated September 29, 2015; February 5, 2016; April 28, 2016; and 
May 19, 2016. Publicly-available versions are in ADAMS under Accession 
Nos. ML15236A265 (Package), ML15272A443, ML16036A091, ML16119A326, and 
ML16141A048, respectively.
    Brief description of amendment request: The amendment would revise 
the Technical Specifications (TSs) by relocating specific surveillance 
frequencies to a licensee-controlled program with the implementation of 
Nuclear Energy Institute document NEI 04-10, ``Risk-Informed Technical 
Specifications Initiative 5b, Risk-Informed Method for Control of 
Surveillance Frequencies'' (ADAMS Accession No. ML071360456). 
Additionally, a new program, the Surveillance Frequency Control 
Program, would be added to TS Section 6, ``Administrative Controls.''
    Date of publication of individual notice in Federal Register: July 
15, 2016 (81 FR 46119).
    Expiration date of individual notice: August 15, 2016 (public 
comments); September 13, 2016 (hearing requests).

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of amendment request: May 16, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16138A247.
    Brief description of amendment request: The amendments would revise 
the Cyber Security Plan implementation schedule for Milestone 8 and 
revise the associated license condition in the Facility Operating 
Licenses.
    Date of publication of individual notice in the Federal Register: 
July 8, 2016 (81 FR 44665).
    Expiration date of individual notice: August 8, 2016 (public 
comments); September 6, 2016 (hearing requests).

IV. Notice of Issuance of Amendments to Facility Operating Licenses and 
Combined Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR chapter I, which are set 
forth in the license amendment.
    A notice of consideration of issuance of amendment to facility 
operating license or combined license, as applicable, proposed no 
significant hazards consideration determination, and opportunity for a 
hearing in connection with these actions, was published in the Federal 
Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation, and/or Environmental Assessment as 
indicated. All of these items can be accessed as described in the 
``Obtaining Information and Submitting Comments'' section of this 
document.

Exelon Generation Company, LLC and PSEG Nuclear LLC, Docket Nos. 50-277 
and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York and 
Lancaster Counties, Pennsylvania

    Date of amendment request: October 2, 2015, as supplemented by 
letter dated March 23, 2016.
    Brief description of amendments: The amendments (1) revised the 
allowable test pressure band in the technical specification (TS) 
surveillance requirements (SRs) for the pump flow testing of the high 
pressure coolant injection system and the reactor core isolation 
system; (2) revised the surveillance frequency requirements for 
verifying the sodium pentaborate enrichment of the standby liquid 
control system; and (3) deleted SRs associated with verifying the 
manual transfer capability of the normal and alternate power supplies 
for certain motor-operated valves associated with the suppression pool 
spray and drywell spray sub-systems of the residual heat removal 
system.

[[Page 50742]]

    Date of issuance: July 5, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendments Nos.: 308 (Unit 2) and 312 (Unit 3). A publicly-
available version is in ADAMS under Accession No. ML16159A148; 
documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Renewed Facility Operating License Nos. DPR-44 and DPR-56: The 
amendments revised the Renewed Facility Operating Licenses and TSs.
    Date of initial notice in Federal Register: December 8, 2015 (80 FR 
76320). The supplemental letter dated March 23, 2016, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 5, 2016.
    No significant hazards consideration comments received: No.

NextEra Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold 
Energy Center, Linn County, Iowa

    Date of amendment request: July 24, 2015.
    Brief description of amendment: The amendment revised Technical 
Specification 1.4, ``Frequency,'' by correcting Example 1.4-1 to be 
consistent with Technical Specifications Task Force (TSTF) Traveler 
TSTF-485, ``Correct Example 1.4-1,'' Revision 0. In addition, the 
amendment revised Example 1.4-5 and Example 1.4-6 to be consistent with 
Amendment No. 258 to the Renewed Facility Operating License.
    Date of issuance: July 13, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 293. A publicly-available version is in ADAMS under 
Accession No. ML15246A408; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. DPR-49: The amendment 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 10, 2015 (80 
FR 69713).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 13, 2016.
    No significant hazards consideration comments received: No.

South Carolina Electric and Gas Company and the South Carolina Public 
Service Authority, Docket Nos. 52-027 and 52-028, Virgil C. Summer 
Nuclear Station (VCSNS), Units 2 and 3, Fairfield County, South 
Carolina

    Date of amendment request: October 1, 2015.
    Brief description of amendment: The amendments consisted of changes 
to the Facility Combined License, Appendix C, ``Inspections, Tests, 
Analyses, and Acceptance Criteria [ITAAC].'' Specifically, the changes 
to the plant-specific Emergency Planning ITAAC removed and replaced 
current references to AP1000 Design Control Document Table 7.5-1, and 
Final Safety Analysis Report (FSAR) Table 7.5-201 on the post-accident 
monitoring system, with references to proposed updated FSAR Table 7.5-1 
in Table C.3.8-1 for ITAAC Numbers C.3.8.01.01.01, C.3.8.01.05.01.05, 
and C.3.8.01.05.02.04.
    Date of issuance: May 2, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment Nos.: 46. A publicly-available version is in ADAMS under 
Package Accession No. ML16074A234. Documents related to these 
amendments are listed in the Safety Evaluation enclosed with the 
amendments.
    Facility Combined License Nos. NPF-93 and NPF-94: Amendments 
revised the Facility Combined Licenses.
    Date of initial notice in Federal Register: November 24, 2015 (80 
FR 73241).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 2, 2016.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, 
Georgia

    Date of amendment request: July 18, 2014, as supplemented by 
letters dated February 27, 2015; May 2, 2016; and June 14, 2016.
    Brief description of amendments: The amendments changed Technical 
Specification 3.9.4, ``Containment Penetrations,'' to allow containment 
penetrations to be un-isolated under administrative controls during 
core alterations or movement of irradiated fuel assemblies within 
containment by adopting a previously NRC-approved Technical 
Specification Task Force (TSTF) Change Traveler TSTF-312, Revision 1, 
``Administratively Control Containment Penetrations.''
    Date of issuance: July 15, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment Nos.: 181 (Unit 1) and 162 (Unit 2). A publicly-available 
version is in ADAMS under Accession No. ML16165A195; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Renewed Facility Operating License Nos. NPF-68 and NPF-81: 
Amendments revised the Renewed Facility Operating Licenses and 
Technical Specifications.
    Date of initial notice in Federal Register: March 3, 2015 (80 FR 
11480). The supplemental letters dated February 27, 2015; May 2, 2016; 
and June 14, 2016, provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not change the staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 15, 2016.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 22nd day of July 2016.

    For the Nuclear Regulatory Commission.
Anne T. Boland,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2016-18290 Filed 8-1-16; 8:45 am]
 BILLING CODE 7590-01-P